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At least 289 records · Page 16

Optimization of boron carbide density using direct current sintering for use as sputtering sources at the National Ignition Facility

Several experimental trials were conducted to determine optimal sintering parameters and compositions for purchased boron carbide powder using spark plasma sintering/direct current sintering. Sputtering targets used to create ablators for the National Ignition Facility have been unable to survive post processing procedures suggesting that in-house fabrication may yield better results due to extended control.

36 MATERIALS SCIENCE↗

SILICON CARBIDE THERMOMETRY USING RAMAN SPECTROSCOPY ON IRRADIATED TRISO PARTICLES

Silicon carbide (SiC) passive thermometry has emerged as a promising post-irradiation examination (PIE) technique for estimating irradiation temperature near the end of irradiation. While dilatometry techniques have been traditionally used to analyze prismatic samples after irradiation, Raman spectroscopy has recently been shown to provide comparable results by analyzing Raman-active phonon modes. In this work, Raman spectroscopy has been applied to the SiC layer of cross-sectioned irradiation tristructural isotropic (TRISO) particles from the AGR-5/6/7 experiment to evaluate the feasibility of particle-scale passive thermometry. Two-dimensional Raman mapping was used to measure the position of the SiC longitudinal optical (LO) phonon, which was then converted to an apparent irradiation temperature using a previously established empirical correlation. Particles from AGR-5/6/7 Compacts 2-2-1 and 5-1-3 were selected as their calculated time-averaged, volume-averaged (TAVA) temperatures (828°C and 706°C, respectively) fall within the range of the sensitivity of the experimental approach. The use of SiC thermometry was anticipated to confirm or highlight potential deviations from calculated end-of-life TAVA temperature across compacts. Four particles from Compact 2-2-1 and two particles from Compact 5-1-3 were selected based on 110mAg inventory (either some measurable activity or below the minimum detection limit) which is commonly used as an indicator of in-pile temperature variation of particles within the same compact. Across every particle selected it was determined that the average LO peak position was located around 968 cm-1 to 969 cm-1. Using the previously determined empirical correlation, this corresponds to an irradiation temperature around 950°C to 966°C, which does not align with the reported TAVA temperature values. This discrepancy in apparent irradiation temperatures likely reflects a combination of uncertainties in the calculated particle temperatures and differences in irradiation history between the present specimens and those used to establish the empirical Raman calibration such as neutron flux (damage rate) and SiC microstructure (as fabricated and irradiated).

Vawdrey, Josh [ORNL]↗

Raman spectroscopy of neutron irradiated silicon carbide

The effects of neutron irradiation on microstructural evolution and the resultant changes in physical and mechanical properties are of critical importance for the development of silicon carbide (SiC) materials for nuclear applications. This study neutron-irradiated βSiC under a wide range of conditions at temperatures between 235 and 750°C and neutron doses of 0.01–11.8 displacements per atom, and then evaluated the effects on the SiC structure using Raman spectroscopy. The SiC optical phonon lines were shifted to lower wavenumbers by irradiation. Correlations were found among the wavenumber of the longitudinal optical phonon line, irradiation-induced swelling, and irradiation temperature. The peak shift also correlated indirectly with decreasing thermal conductivity of irradiated SiC. The irradiation-induced peak shift is explained by combinations of lattice strain, reduction of the elastic modulus, and other factors including decreasing coherent domain size. These findings bridge irradiation-induced microstructural changes and property changes and illustrate how Raman spectroscopy is a useful tool for nondestructively assessing irradiated SiC materials for nuclear applications.

Koyanagi, T.↗

Using a carbon vacancy reduction material to increase average carrier lifetime in a silicon carbide semiconductor device

A semiconductor die and a process for fabricating the semiconductor die are disclosed. The semiconductor die has a substrate and a silicon carbide (SiC) epitaxial structure on the substrate. The SiC epitaxial structure includes at least a first N-type SiC layer, at least a first P-type SiC layer, and carbon vacancy reduction material, which has been implanted into a surface of the SiC epitaxial structure. Further, the SiC epitaxial structure has been annealed to mobilize the carbon vacancy reduction material to diffuse carbon atoms substantially throughout the SiC epitaxial structure, thereby increasing an average carrier lifetime in the SiC epitaxial structure.

36 MATERIALS SCIENCE↗

Silicon carbide power inverter/rectifier for electric machines

The present disclosure involves a two stage inverter, a system for electrical power conversation, and a method of converting electrical power using silicon carbide (SiC) metal-oxide-semiconductor field-effect transistors (MOSFETs). One example implementation includes using two or more SiC MOSFETs in series with each MOSFET having a gate terminal for triggering a state switch between an on (conducting) and off (non-conducting) state of the MOSFET. An AC terminal is connected between the series SiC MOSFETS, and the series SiC MOSFETs are connected across a DC bus and in parallel with one or more capacitors.

Shenoy, Suratkal P.↗

Fission Product Diffusion in Silicon Carbide: As Revealed by Computer Simulations and Experimental Measurements

The diffusion and release of fission products through silicon carbide in the tri-structural isotropic (TRISO) fuel particles remain unsolved for decades. The underlying mechanism is quite challenging to be determined. To help unveil the mysterious story, the current work applies molecular dynamics method to show the stability of silver, palladium, ruthenium and iodine as an interstitial and their atomic diffusion along coincident site lattice (CSL) boundary, especially ?3 grain boundary (GB). The major finding presents a much faster diffusion along GB than in bulk for all elements considered. The reasonably close estimate to experiments and simulations where available has confirmed the important role of grain boundary diffusion of Ag and Pd in SiC. However, the discrepancy addressed in Ag with measurements from fuel studies suggest a more complicated mechanism, which might be in correlation with high energy grain boundaries or the presence of crack. The subsequent characterization of Ru and I distribution in SiC-PyC-SiC diffusion couples, which have been ion irradiated at 900?C to 10 dpa and 20 dpa, has performed by secondary ion mass spectrometry (SIMS) analysis. The experiment measurements correlate well with the grain boundary diffusion by simulation, which provide further evidences that the grain boundary diffusion cannot be neglected once the fission products are accessible at the grain boundary.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Silicon Carbide Multilayer Piping for High Temperature sCO 2 Brayton Cycle

Efficiencies of greater than 50% in supercritical carbon dioxide (sCO 2 ) Brayton power cycle systems can be achieved only at turbine inlet temperatures of above 700°C. In support of the push to higher temperatures, a finite element model was developed by Materials Research & Design, Inc. (MR&D) with support from Ceramic Tubular Products, LLC. (CTP) to guide the refinement of CTP’s high temperature ceramic multilayer piping. The multilayer technology combines the advantages of a monolithic silicon carbide (SiC) tube and a SiOC f /SiOC ceramic matrix composite (CMC), the result of which is a material with high-temperature strength and stability, high mechanical and thermal shock resistance, and high corrosion resistance. In addition to fabricating test specimens to refine the finite element model, long-duration, high temperature CO 2 exposure tests were performed by Sandia National Laboratories (SNL) on two varieties of inner monolithic SiC.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Integrated silicon carbide diode rectifier circuits

An integrated silicon carbide rectifier circuit with an on chip isolation diode. The isolation diode can be a channel-to-substrate isolation diode or a channel to channel isolation diode. the circuit teaches an integrated diode rectification circuit for use with a two phase center tap transformer having a first voltage output, a second voltage output, and a center tap output with a single chip having a first half-wave rectifier connected to the first voltage output, a second half-wave rectifier connected to the second voltage output, and a floating substrate connection connected to the center tap output and an on chip first channel-to-substrate isolation diode electrically connected between the first half-wave rectifier and the floating substrate.

Barlow, Matthew↗

Ab Initio Evaluation of Uranium Carbide S(α,β) and Thermal Neutron Cross Sections

Uranium Carbide (UC) is a nuclear fuel material which offers better neutron economy and lower fuel-cycle costs compared to conventional mixed-oxide fuels. UC’s lattice binding and dynamical properties impact thermal neutron scattering and low temperature epithermal resonance absorption. The Thermal Scattering Law (TSL) describes the scattering system available energy and momentum transfer states. There is no TSL evaluation for UC in the ENDF/B-VIII.0 database; herein, ab-initio lattice dynamics (AILD) techniques are invoked to calculate the phonon spectrum for UC using spin-orbit-coupling density functional theory (DFT). The TSLs, inelastic and elastic thermal scattering cross sections for Uranium and Carbon in UC, respectively, are calculated in FLASSH for use in higher fidelity reactor design calculations.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Ultra-high temperature carbide foams and methods of fabricating the same

Ultra-high temperature carbide (UHTC) foams and methods of fabricating and using the same are provided. The UHTC foams are produced in a three-step process, including UHTC slurry preparation, freeze-drying, and spark plasma sintering (SPS). The fabrication methods allow for the production of any kind of single- or multi-component UHTC foam, while also providing flexibility in the shape and size of the UHTC foams to produce near-net-shape components.

Agarwal, Arvind↗

Carbide-based fuel assembly for thermal propulsion applications

Carbide-based fuel assembly includes outer structural member of ceramic matrix composite material, the interior surface of which is lined in higher temperature regions with an insulation layer of porous refractory ceramic material. Continuous insulation layer extends the length of the fuel assembly or separate insulation layer sections have a thickness increasing step-wise along the length of the fuel assembly from upper (inlet) section towards bottom (outlet) section. A fuel element positioned inward of the insulation layer and between support meshes has a fuel composition including HALEU and the form of a plurality of individual elongated fuel bodies or one or more fuel monolith bodies containing coolant flow channels. Fuel assemblies are distributively arranged in a moderator block, with upper end of the outer structural member attached to an inlet for propellant and lower end of the outer structural member operatively interfaced with a nozzle forming a nuclear thermal propulsion reactor.

Barringer, Eric A.↗

Ultra-high temperature carbide foams and methods of fabricating the same

Ultra-high temperature carbide (UHTC) foams and methods of fabricating and using the same are provided. The UHTC foams are produced in a three-step process, including UHTC slurry preparation, freeze-drying, and spark plasma sintering (SPS). The fabrication methods allow for the production of any kind of single- or multi-component UHTC foam, while also providing flexibility in the shape and size of the UHTC foams to produce near-net-shape components.

Agarwal, Arvind↗

Micro-Tensile Properties of Irradiated AGR-2 TRISO Fuel Pyrolytic Carbon (PyC) and Silicon Carbide (SiC) Coatings

Tristructural isotropic (TRISO) coated nuclear fuel particles are emerging as a versatile option for new reactor designs, with the silicon carbide (SiC) layer crucial for retaining fission products. However, the mechanical properties of TRISO coating layers, particularly after irradiation, are not fully understood due to their small size and high radioactivity. Recent in situ micro-tensile testing of various TRISO layers aims to better understand the SiC layer's failure mechanisms, advancing TRISO fuel qualification. These micro-tensile results will be presented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Thermal Optimization and Experimental Demonstration of a Silicon Carbide, Half-Bridge Power Module

This paper describes modeling and experimental work to design and demonstrate a silicon carbide (SiC), single-side-cooled power module. Two SiC metal-oxide-semiconductor field-effect transistors (MOSFET) were used for each switch position and packaged on a metalized ceramic substrate. Thermal modeling was first conducted to optimize the module packaging to minimize junction-to-fluid thermal resistance and device temperature variations. This procedure identified the optimal device placement and baseplate thickness to maximize thermal performance. A heat exchanger - incorporating impinging jets of water-ethylene glycol - was designed to cool the module's baseplate to minimize thermal resistance and pumping power. The heat exchanger used relatively low fluid velocities (<1.5 m/s) to minimize erosion-corrosion effects and relatively large fluid channel sizes (>=1 mm) to prevent clogging. The power module concept was fabricated, and experiments were conducted to measure its thermal performance. Direct current was used to power the module devices, and junction temperatures were measured via device temperature-sensitive parameters. The experimental results were found to be in good agreement with model predictions (within 2%) and yielded junction-to-fluid thermal resistance values as low as about 15 mm2 K/W at relatively low pumping power values. The results demonstrate the improved performance associated with jet impingement cooling and represent an improvement over existing channel-flow-type cold plates, which can enable improvements to power density and reliability.

36 MATERIALS SCIENCE↗

Multiscale, mechanistic calculation of the effective silver diffusion coefficient in polycrystalline silicon carbide: application to silver release in AGR-1 TRISO particles

The silicon carbide (SiC) layer in tristructural isotropic (TRISO) fuel particles serves as a barrier to prevent the escape of fission from the fuel kernel. The release of silver (Ag) is a concern due to the long half-life of the 110mAg isotope. In this study, the effective diffusion coefficient of the fission product Ag through the grain boundary (GB) network is calculated using a combination of atomistic and phase-field methods. Atomistic calculations of Ag diffusivity in SiC bulk and GBs are leveraged to develop a mesoscale effective Ag diffusion coefficient (Deff) in SiC. Since GBs serve as pathways for Ag diffusion, Deff is defined as a function of temperature, microstructure variables, and fluence. Deff is implemented in the fuel performance code Bison to predict Ag release from AGR-1 TRISO fuel particles. We hereby quantify the impact of SiC grain size and irradiation on Ag release and improve Bison's predictions.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Understanding the effect of irradiation and silicon carbide microstructure on silver diffusion in TRISO particles

The silicon carbide (SiC) layer in tristructural isotropic (TRISO) fuel particles serves as a barrier to prevent the escape of fission from the fuel kernel. The release of silver (Ag) is a concern due to the long half-life of the 110mAg isotope. In this study, a combination of atomistic and phase-field methods focuses on quantifying the effect of irradiation and SiC microstructure on Ag diffusion. Atomistic calculations provide the Ag diffusivity in SiC bulk and GBs, also providing an estimate of the effect of irradiation defects on Ag diffusivity in bulk SiC. These lower length scale efforts are leveraged to develop a mesoscale effective Ag diffusion coefficient (Deff) in SiC. Deff is defined as a function of temperature, microstructure variables, and fluence. Deff is implemented in the fuel performance code Bison to predict Ag release from AGR-1 and AGR-2 TRISO fuel particles. We hereby quantify the impact of SiC grain size and irradiation on Ag release and improve Bison's predictions.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Strength, deformation, and equation of state of tungsten carbide to 66 GPa

Here, strength, texture, and equation of state of hexagonal tungsten monocarbide (WC) have been determined under quasi-hydrostatic and non-hydrostatic compression to 66 GPa using angle-dispersive X-ray diffraction in the diamond anvil cell. Quasi-hydrostatic compression in a Ne pressure medium demonstrates that nanocrystalline WC is slightly less incompressible than bulk-scale WC, with respective bulk moduli of K 0 = 377 ± 7 and 397 ± 7 GPa and pressure derivatives K 0 ’ = 3.8 ± 0.3 and 3.7 ±0.3. This decrease in incompressibility with grain size is similar to behavior observed in other ceramics. Under nonhydrostatic compression, WC supports a mean differential stress of ~12-15 GPa at plastic yielding, which occurs at ~30 GPa. Strength in WC is anisotropic, with the (001) plane supporting 29-42% higher stress than stresses calculated from mean strain. Simulations using an Elasto-ViscoPlastic Self-Consistent model indicate that strength inferred from lattice strain theory may be overestimated due to effects of plastic deformation. Plastic deformation generates a texture maximum near $\langle\bar{2}110\rangle$ in the compression orientation, initially through prismatic slip on the {$10\bar{1}0$}$\langle\bar{1}2\bar{1}0\rangle$ and {$10\bar{1}0$}$\langle0001\rangle$ slip systems, followed by activation of pyramidal slip on {$10\bar{1}1$}$\langle\bar{2}113\rangle$ at ~40-50 GPa.

36 MATERIALS SCIENCE↗